The present invention relates to pressurized water nuclear reactors (PWRs). More particularly, the present invention relates to an incore instrumentation system for a pressurized water nuclear reactor.
Before discussing the invention per se, some background information relating to PWRs is presented. For further background, see "Nuclear Power Plant Systems and Equipment," by K. C. Lish, ISBN 0-8311-1078-3.
A pressurized water nuclear power station contains a closed loop of pressurized water which removes heat energy from the core of the PWR, and transfers the energy to a second system for generating steam. The steam, in turn, drives a turbine generator which produces electrical power.
The reactor system consists of a pressure vessel containing the nuclear fuel which generates the heat energy, a steam generator in which the heat energy is used to generate steam, a circulating pump which circulates the coolant, and a pressurizer that maintains and controls system pressure. A nuclear island of a PWR power station consists of three or four buildings, depending on the individual design. The three buildings that must always be present are the reactor containment building, the fuel handling building, and the auxiliary building. The fourth, optional, building, is the control building. The space provided in this last building is always required, but some designs include it as an additional wing to another building.
The reactor building is typically a spherical structure or cylindrical structure with a hemispherical or elliptical top head. The bottom head may be either of the two shapes just mentioned or a flat head, depending on the structural design. The reactor building contains all of the primary system equipment. It also has a polar crane which is used during normal refueling, in-service inspection, and for general maintenance.
The fuel handling building contains the spent fuel storage pool, a bridge crane over the pool for handling the fuel shipping cask, and all of the spent fuel cooling system equipment. This building is connected to the reactor building by a fuel transfer tube.
The control building, or wing, contains the central control room with its console and control panels, as well as the relay room with its relays and controllers. The entire plant is controlled from this spot.
FIG. 1 depicts a typical PWR assembly 10. The reactor assembly consists of a pressure vessel 12 with a thermal shield, core support plates 32, 54, control rods 22, etc. Coolant enters the vessel and flows down the annulus between the core barrel and the vessel wall; in the process it cools the thermal shield. The coolant turns, flows up through the fuel elements and out of the vessel to the steam generator. Control rods enter through the top head 18 and the control rod drive mechanisms are mounted on the top head. The drive mechanisms and top head are handled as one package during plant refueling. This package is often referred to as an "integrated head package," or "IHP."
In FIG. 1, the neutron detectors 36 entering the bottom head are combined with thermocouples which measure coolant temperature at the outlet of an individual fuel assembly 30. Together they are known as "incore instrumentation." They are information instrumentation only and provide data with which to calculate the power produced in the different regions of the core.
FIG. 2 depicts a known arrangement of an instrumentation system for a PWR. In this embodiment, the instrumentation system includes a guide tube cluster assembly 46 and a number of guide tubes 48 penetrating an upper support plate 42a. The primary difference between this arrangement and the arrangement shown in FIG. 1 is that this arrangement employs flexible guide tubes 48 and cluster assembly 46, in contrast to the rigid guide tubes shown in FIG. 1.
Returning to FIG. 1, during the refueling operation all internals inside the core barrel above the upper core plate 54 are removed. This permits direct access to the top of the fuel assemblies 30 for manipulation.
The fuel rods are assembled into grids in, e.g., typically a 17.times.17-square array. One assemblage of fuel rods into grids is called a "fuel element assembly," and is the smallest fuel unit handled in a power station. A core is built up by grouping fuel assemblies 30 side-by-side in an essentially cylindrical shape. The higher the power level desired, the more fuel element assemblies are used. Core diameters typically run from 9 ft. to 13 ft., depending on the power level. Core height is typically 10 to 14 ft. for large PWRs.
The control rods are cylindrical rods located inside control-rod guide channels 20 in various locations in a fuel element assembly, in place of fuel rods. The different manufacturers do not use the same neutron absorber nor do they have the same number of rods.
The control rod drives 14 are mechanical, and are either infinitely positionable or move in a series of discrete steps. All rod drives, regardless of manufacturer, incorporate a magnetic latch device which is de-energized for "scram" (i.e., emergency shutdown). The control rods withdraw upward and drop into the core by gravity during scram. The control rod drives enter the reactor pressure vessel through the top head 18.
When the reactor is operating, measurements of neutron flux must be carried out inside the reactor core, in various places distributed along its cross-section and along its height. These measurements are carried out by the incore nuclear instrumentation, which comprises the subject matter of the present invention.
An instrumentation device for the core of a nuclear reactor, in which the measurement conduits pass through the closure head is described in French Patent 2,065,512. The instrumentation device described therein comprises guide tubes capable of receiving either neutron detectors in a fixed position or detectors which can move inside the respective tubes. The guide tubes pass through the vessel closure head inside columns and are then distributed along the cross-section of the core by beam-shaped support arms arranged under the vessel closure head, which enable each of the director tubes to be directed towards a guide tube of a particular core fuel assembly.
With this prior art device, in order to recharge the core, after the vessel has been depressurized, the devices permitting the leakproof passage of the instrumentation columns are disassembled and then the vessel closure head is disassembled to permit access to the support arms and the director tubes.
There are several disadvantages with this prior art device, however. Firstly, the withdrawal of the support arms, each carrying a plurality of guide tubes independently of one another, must be performed with the aid of a special tool, so as to preserve the shape and the distribution of these conduits. This requires numerous handling operations requiring special tools. The time involved can therefore be very long, which increases the length of reactor downtime. This can be a major disadvantage. Secondly, in order to avoid having a large number of support arms of complex shape, the number of assemblies into which a neutron flux detector can be introduced is limited, at the expense of the accuracy of the flux pattern measurements in the core.
In the above case, as in all cases, the measurement conduits comprise an end part which is introduced into the core and a second, opposed end part which enters a measurement room arranged in the vicinity of the reactor core. The measurement conduits are mounted slideably in the instrumentation tubes and may be withdrawn or installed merely by pulling or pushing on their ends, from within the measurement room. The measurement conduits must, in fact, be withdrawn from the fuel assemblies of the core when the reactor core is being recharged. Moreover, the measurement conduits must enter the reactor vessel through leakproof passages through which guide tubes of great length are connected, linking the reactor vessel with the measurement room.
In a second embodiment which is known and widely employed, the passages for the measurement conduits consist of fittings provided on the convex bottom of the reactor vessel. The instrumentation guide tubes connected to these fittings have a path in the shape of an arc of a circle of large radius of curvature, joining the bottom of the vessel to the measurement room. The provision of fittings in the convex bottom of the vessel complicates the manufacture of this component and leads to difficulties in the observance of safety standards. Moreover, the structure of the reactor building must be designed to permit the passage of instrumentation guide tubes of great length along a trajectory which is a circular arc. As a result of this, the design and the construction of the reactor building are made more difficult and costly.
Furthermore, access of the fittings passing through the bottom of the vessel is very difficult, and this complicates the operations of monitoring these fittings in order to ensure that the reactor operates in complete safety. In addition, the instrumentation guide tubes connected to the bottom of the vessel are always filled with the core coolant water, with the result that this water can enter the measurement room in the event of a fault in the sealing of the guide tube.
An instrumentation device has also been proposed and used, in which the measurement conduits pass through the vessel closure head. See, e.g., French Patent 2,065,512. This avoids the disadvantages associated with the need to provide fittings in the bottom of the vessel and a complex reactor building structure to permit the passage of the instrumentation guide tubes towards the measurement room. However, in a layout of this kind, a part of the instrumentation device called "upper internal equipment" is carried directly by the closure head, which complicates the operation of dismantling the closure head, as well as the instrumentation handling and storage operations associated with the closure head during reactor stoppages.
During a reactor stoppage for recharging and maintenance, the upper internal equipment is withdrawn and arranged on the storage stand in the reactor pool. In the case of a core instrumentation device comprising guide tubes or measurement conduits passing through the closure head, the known arrangements do not make it possible to perform the handling of the upper internal equipment in a simple and quick manner. The advantages obtained in respect of the design and construction of the reactor, when the core instrumentation passes through the vessel closure head, are therefore accompanied by very considerable disadvantages insofar as the use and maintenance of the reactor are concerned. For this reason, most of the nuclear reactors which are now in service incorporate instrumentation tubes which pass through the bottom of the vessel.
In addition, as a result of the Three Mile Island incident in 1979, there has been an emergence of DOE, EPRI and NRC preferences for incore instrumentation systems which do not penetrate the reactor vessel (RV) below the core, and which employ fixed incore detectors (FIDs). These factors are believed to improve plant safety and operational control.
In accordance with the above discussion, an object of the present invention is to provide a simplified incore instrumentation system (SICIS) for a PWR that eliminates bottom reactor vessel penetrations and also employs FIDs for measuring core power distribution. A further object of the invention is to provide a SICIS that is easily and safely installed in and withdrawn from the PWR. These and other objectives are achieved by the present invention.